Field of the Invention
The present invention relates generally to fuel assemblies for a nuclear reactor and, more particularly, is concerned with an improved fuel rod for a boiling water reactor (BWR) fuel assembly which implements criterion for enhanced transient response and system stability.
Description of the Prior Art
Typically, large amounts of energy are released through nuclear fission in a nuclear reactor with the energy being dissipated as heat in the elongated fuel elements or rods of the reactor. The heat is commonly removed by passing a coolant in the heat exchange relation to the fuel rods so that the heat can then be extracted from the coolant to perform useful work.
In nuclear reactors generally, a plurality of the fuel rods are grouped together to form a fuel assembly. A number of such fuel assemblies are typically arranged in a matrix to form a nuclear reactor core capable of self-sustained, nuclear fission reaction. The core is submerged in a flowing liquid, such as light water, that serves as the coolant for removing heat from the fuel rods and a neutron moderator.
A typical fuel rod is composed of an elongated cladding tube having end closures or plugs attached to opposite ends of the tube so as to provide a hermetically sealed chamber in which a plurality of nuclear fuel pellets are disposed end-to-end in a stack. Both the tube and pellets are preferably circular in cross-section and the pellets have a length approximately twice their diameter. The diameter of a pellet is slightly less than that of the tube so that a clearance space or gap is formed therebetween to accommodate any swelling of the pellet during operation of the reactor.
The trend in fuel rod design in all types of nuclear reactors has been to make the thickness of the fuel rod cladding tube as thin as possible for various reasons. For example, in U.S. Pat. No. 3,376,201 to Bain, it is mentioned that the protective sheath (or cladding tube) will, undesirably, absorb neutrons and must be made as thin as possible consistent with the aim of preventing distortion of the fuel element (or rod). With respect to a BWR, such trend overlooks the profound impact that fuel rod design has on nuclear coupled/decoupled thermal hydraulic transients which take place in the reactor. A nuclear coupled transient is one where the thermal hydraulic behavior is coupled to neutronic feedback due to the void reactivity feedback mechanism, whereas, a decoupled transient implies one where this void reactivity feedback is neglected. These transients refer to fuel rod responses to key pertubations, for example, neutron flux/power response to pertubations in the flow rate system pressure, subcooling, etc.
Consequently, a need exists for redirection of attention to gaining a systematic understanding of the effect of fuel rod dynamics on BWR system transient/stability response, with an eye toward modification of the conventional fuel rod design to improve such response.